ML14041A093

Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Revised Pressure-Temperature Limits (TAC Nos. MF0763, MF0764, MF0765)

ML14041A093
Person / Time
Site:	Oconee
Issue date:	02/27/2014
From:	Guzman R V Plant Licensing Branch II
To:	Batson S Duke Energy Carolinas
Guzman R V
References
TAC MF0763, TAC MF0764, TAC MF0765
Download: ML14041A093 (34)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Scott Batson Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752 February 27, 2014

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3, ISSUANCE OF AMENDMENTS REGARDING REVISED PRESSURE-TEMPERATURE LIMITS (TAC NOS. MF0763, MF0764, AND MF0765)

Dear Mr. Batson:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 384, 386, and 385 to Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station (ONS), Units 1, 2, and 3, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated February 22, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13058A059), as supplemented by letters dated September 10 (ADAMS Accession No. ML 13259A 120), October 25 (ADAMS Accession No. ML 13305A 120), November 29 (ADAMS Accession No. ML 13337 A 169), and December 16, 2013 (ADAMS Accession No. ML 13350A098). These amendments revise the ONS, Units 1, 2, and 3 TSs by replacing the reactor pressure vessel pressure-temperature (P-T) limits in TS 3.4.3 with new P-T limits applicable to 54 effective full power years. In addition, the amendments change the operational requirements for unit heatup and cooldown in TS Tables 3.4.3-1 and 3.4.3-2. A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

S. Batson If you have any questions, please call me at 301-415-1030. Docket Nos. 50-269, 50-270, and 50-287

Enclosures:

1. Amendment No. 384 to DPR-38 2. Amendment No. 386 to DPR-47 3. Amendment No. 385 to DPR-55 4. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 384 Renewed License No. DPR-38 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. DPR-38, filed by Duke Energy Carolinas, LLC (the licensee), dated February 22, 2013, as supplemented by letters dated September 10, October 25, November 29, and December 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 1 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 384, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-38 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 27, 2014 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 386 Renewed License No. DPR-47 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated February 22, 2013, as supplemented by letters dated September 10, October 25, November 29, and December 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 2 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 386, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-47 and the Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 27, 2014 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION. UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 385 Renewed License No. DPR-55 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), dated February 22, 2013, as supplemented by letters dated September 10, October 25, November 29, and December 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 3 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 385, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance. *

Attachment:

Changes to Renewed Facility Operating License No. DPR-55 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 27, 2014 ATTACHMENT TO LICENSE AMENDMENT NO. 384 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 386 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 385 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Pages Licenses License No. DPR-38, page 3 License No. DPR-47, page 3 License No. DPR-55, page 3 TSs Page 3.4.3-3 Page 3.4.3-4 Page 3.4.3-5 Page 3.4.3-6 Page 3.4.3-7 Page 3.4.3-8 Page 3.4.3-9 Page 3.4.3-10 Page 3.4.3-11 Page 3.4.3-12 Page 3.4.3-13 Insert Pages Licenses License No. DPR-38, page 3 License No. DPR-47, page 3 License No. DPR-55, page 3 TSs Page 3.4.3-3 Page 3.4.3-4 Page 3.4.3-5 Page 3.4.3-6 Page 3.4.3-7 Page 3.4.3-8 Page 3.4.3-9 Page 3.4.3-10 Page 3.4.3-11 Page 3.4.3-12 Page 3.4.3-13 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 384 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1l1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-38 Amendment No. 384 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 386 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 386 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 385 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following antitrust conditions: Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity. Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction. 1. As used herein: (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another. (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 385 CONSTRAINT RC Temperatur-e<a> RC Pumps Table3.4.3-1 (page 1 of1) Operational Requirements for Unit Heatup RC TEMPERATURE<a> HEATUP RATE T < 270°F s 30°F in any % hr period T 270°F s 50°F in any % hr period T<100°F NA 1 00°F s T < 300°F NA T 300°F NA RCS PIT Limits 3.4.3 ALLOWED PUMP COMBINATION NA NA No pumps s two pumps Any (a) RC Temperature is cold leg temperature if one or more RC pumps are in operation; otherwise it is the LPI cooler outlet temperature. OCONEE UNITS 1, 2, & 3 3.4.3-3 Amendment Nos. 384, 386, & 385 CONSTRAINT RC Temperature's) RC Pumps Table 3.4.3-2 (page 1 of 1) Operational Requirements for Unit Cooldown RC TEMPERATURE(a> T 2': 270°F 140°F:::; T < 270°F T< 140°F RCS depressurized(c) T 2': 300°F 1 00°F s; T < 300°F T < 100°F COOLDOWN RATE(bl :::; 50°F in any 1/2 hour period :::; 25°F in any 1/2 hour period :::; 50°F in any one hour period :::; 50°F in any one hour period NA NA NA RCS PIT Limits 3.4.3 ALLOWED PUMP COMBINATION NA NA NA NA Any ::;two pumps No pumps (a) RC Temperature is cold leg temperature if one or more RC pumps are in operation or if on natural circulation cooldown; otherwise it is the LPI cooler outlet temperature. (b) These rate limits must be applied to the change in temperature indication from cold leg temperature to LPI cooler outlet temperature per Note (a). (c) When the RCS is depressurized such that all three of the following conditions exist: a) RCS temperature< 200°F, b) RCS pressure< 50 psig, c) All RC Pumps off, the maximum cooldown rate shall be relaxed 50°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. OCONEE UNITS 1, 2, & 3 3.4.3-4 Amendment Nos. 384, 386, & 385 I Cl 'iii c. :I Ill Ill Cl) ... a.. en 0 0:: "C ca "C = 2400 2000 j Heatup Temp. Press. ' ("F) (psig) I 6o 528 i 105 528 1600 115 545 120 553 135 557 170 557 170 729 1200 175 804 190 903 210 1089 800 400 0 0 50 Criticality Limit Temp. Press. ("F) (psig) 253 0 253 1124 270 1364 290 1771 310 2374 -composite HU Curve --Criticality Limit P-T Limit Curve i i ' 100 150 200 , /1 li RCS PIT Limits 3.4.3 ):--t----------1 i I ," I _____ , :, i I* I II I r -r-----'1 ,I *-t----:. :, 250 300 350 Indicated RCS Inlet Temperature, oF The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-1 (page 1 of 1) RCS Normal Operational Heatup Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 1 OCONEE UNITS 1, 2, & 3 3.4.3-5 Amendment Nos. 384, 386, & 385 C) 'iii Q. a) ... 1/) 1/) (I) ... a.. en 0 0:: '0 (I) -"' '0 c: RCS PIT Limits 3.4.3 2400 -------------*-*-*---***-*-* Normal Cool down Temp. Press. -composite CD Curve (oF) (psi g) 2000 . 251 2231 ----- 246 2221 231 1779 211 1363 1600 . 191 1083 190 1053 186 1029 181 981 171 837 1200 166 824 161 765 155 710 146 636 800 135 611 110 531 105 527 100 513 400 70 513 65 512 60 506 0 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, oF The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-2 (page 1 of 1) RCS Normal Operational Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 1 OCONEE UNITS 1, 2, & 3 3.4.3-6 Amendment Nos. 384, 386, & 385 0 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-3 (page 1 of 1) RCS Leak and Hydrostatic Test Heatup and Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 1 OCONEE UNITS 1, 2, & 3 3.4.3-7 Amendment Nos. 384, 386, & 385 C) 'iii Q. e ::1 Ill Ill f c.. U) () IX: "C 2 cu .!::! "C .E RCS PIT Limits 3.4.3 2400 . -composite HU Curve ,, I! --Criticality Limit P-T Limit Curve I I 2000 I ,-r--Normal Heatup Criticality Limit 1 Temp. Press. Temp. Press. I I I (oF) (psig) (oF) (psig) I

• I 60 527 243 0 I I 1600 105 527 243 1126 115 543 265 1446 I tl 120 550 280 1764 I I 170 550 300 2365 I, 1200 170 776 --------J---185 914 ti 205 1145 I 225 1446 I 800 240 1764 I 260 2365 I I I

• 400 .-+---I' I I 0 0 50 100 150 200 250 300 350 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-4 (page 1 of 1) RCS Normal Operational Heatup Limitations Applicable for the First 54 EFPY-Oconee Nuclear Station Unit 2 OCONEE UNITS 1, 2, & 3 3.4.3-8 Amendment Nos. 384, 386, & 385

.!2> Ul c. a) ... :::J Ul Ul Q.. t/) 0 0:: '0 .s cu (J '0 c: RCS PIT Limits 3.4.3 2400 Normal Cooldown Temp. Press. -composite CD Curve (oF) (psig) 2000 251 2365 241 2365 231 2039 211 1539 1600 191 1203 190 1166 186 1138 176 1021 171 907 1200 ! . ----------166 907 161 871 155 802 800 146 708 ----___ [ 135 670 i 110 557 105 557 100 544 400 75 544 70 534 60 518 0 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-5 (page 1 of 1) RCS Normal Operational Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 2 OCONEE UNITS 1, 2, & 3 3.4.3-9 Amendment Nos. 384, 386, & 385

.21 f/) Q. i f/) f/) 0.. (/) (.) 0:: '0 s cu .!::! '0 r:: RCS P/T Limits 3.4.3 2400 -ISLH Composite (HU/CD) Curve 2000 ISLH Composite I t-Temp. Press. (*F) (psig) 60 557 1600 -------------------------lOS 557 110 614 115 753 120 763 1200 170 763 170 1065 190 1316 210 1643 800 225 1958 240 2382 400 0 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-6 (page 1 of 1) RCS Leak and Hydrostatic Test Heatup and Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 2 OCONEE UNITS 1, 2, & 3 3.4.3-10 Amendment Nos. 384, 386, & 385

-composite HU --Criticality Limit p. T Limit Curve RCS PIT Limits 3.4.3 1 I i I I 2000 ...... r----:----'---Cr-it-ica-:-li_ty_,_Li_m-:-itl-*******-******-**-*l-**-*******-***-**-**-**-**-***f*-***-*******-**-**-***-******i*-*******-**-**-**l*-*-***-j"***-*-*,.-**-**-**-*-** Normal Heatup Temp. Press. I I (°F) (psig) I e 275 o I Temp. Press. (oF) (psig) 60 470 275 1148 *-** -**-------}*-*-------*--295 1453 Jl 1600 *-* 170 470 170 638 310 1772 I I 325 2081 I I 335 2364 *1 . -**1** I 1200 190 746 210 907 220 990 235 1150 255 1453 270 1772 I , ... 1 ............ I 800 275 1856 285 2081 295 2346 400 i**************--*********+*-*****-*** 0 ______ 0 50 100 150 200 250 300 350 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-7 (page 1 of 1) RCS Normal Operational Heatup Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 3 OCONEE UNITS 1, 2, & 3 3.4.3-11 Amendment Nos. 384, 386, & 385 2400 2000 0) *;n c. 1600 ::;, U) U) (I) ... ll. C/) 1200 (.) 0:::: "C (I) 800 -C\'S (J :s .E 400 0 RCS PIT Limits 3.4.3 .. **r*********** L __ Normal Cooldown Temp. Press. -composite CD Curve (oF) (psig) 270 2359 t 255 1895 251 1675 246 1668 , 241 1559 I 231 1371 211 1091 206 1037 201 971 195 928 /v 190 860 171 701 166 642 ' 156 582 i v 146 533 135 529 110 487 1-105 486 100 476 60 476 ! 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-8 (page 1 of 1) RCS Normal Operational Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 3 OCONEE UNITS 1, 2, & 3 3.4.3-12 Amendment Nos. 384, 386, & 385 Cl *u; Q. o) ... :::3 Ill Ill (I) ... Q. tJ) (.) 0:: "0 (I) 1U (J :s c RCS PIT Limits 3.4.3 2400 2000 ----ISLH Composite Temp. Press. (oF) (psi g) 60 557 1600 105 557 110 614 115 655 170 655 170 880 1200 -**-" 190 1024 210 1238 225 1414 240 1649 800 255 1966 270 2392 400 0 0 50 100 150 200 250 300 Indicated RCS Inlet Temperature, °F The regions of acceptable operation are below and to the right of the limit curves. Margins are included for the pressure differential between point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. Margins for instrument error are not included. Note: Heatup and Cooldown rate restrictions and Reactor Coolant Pump combination restrictions during Heatup and Cooldown are required, as identified in text. Figure 3.4.3-9 (page 1 of 1) RCS Leak and Hydrostatic Test Heatup and Cooldown Limitations Applicable for the First 54 EFPY -Oconee Nuclear Station Unit 3 OCONEE UNITS 1, 2, & 3 3.4.3-13 Amendment Nos. 384, 386, & 385 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 384 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 386 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-47 AND AMENDMENT NO. 385 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION. UNITS 1, 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287

1.0 INTRODUCTION

By application dated February 22, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13058A059), as supplemented by letters dated September 10 (ADAMS Accession No. ML 13259A 120), October 25 (ADAMS Accession No. ML 13305A 120), November 29 (ADAMS Accession No. ML 13337 A 169), and December 16, 2013 (ADAMS Accession No. ML 13350A098), Duke Energy Carolinas, LLC (Duke Energy, the licensee), submitted a license amendment request (LAR) to replace the current reactor pressure vessel (RPV) pressure-temperature (P-T) limits in Technical Specifications (TSs) 3.4.3, with new P-T limits applicable to 54 effective full power years (EFPYs) for Oconee Nuclear Station (ONS), Units 1, 2, and 3. In addition, the LAR proposed to change the operational requirements for unit heatup and cooldown in TS Tables 3.4.3-1 and 3.4.3-2. The licensee revised the P-T limits based on BAW-10046A, Revision 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of [Part 50 of Title 10 of the Code of Federal Regulations (1 0 CFR Part 50)], Appendix G," which is supplemented by an approved exemption to use the alternative initial nil-ductility transition reference temperature (RT NoT) in BAW-2308, Revisions 1-A and 2-A, "Initial RTNoTof Linde 80 Weld Materials." The supplemental letters dated September 10, October 25, November 29, and December 16, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 16, 2013 (78 FR 22568). Enclosure 4

2.0 REGULATORY EVALUATION

The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (1 0 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The NRC staff evaluates the acceptability of a facility's proposed P-T limits based on the following NRC regulations and guidance: Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50; Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials"; Generic Letter (GL} 92-01, Rev. 1, "Reactor Vessel Structural Integrity"; GL 92-01, Rev. 1, Supplement 1, "Reactor Vessel Structural Integrity"; and Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock." Appendix G to 10 CFR Part 50 requires that RPV beltline materials must have upper-shelf energy (USE) of no less than 75 ft-lb initially and 50 ft-lb throughout the life of the vessel. It also requires that facility P-T limit curves for the RPV be at least as conservative as those obtained by applying the linear elastic fracture mechanics (LEFM) methodology of Appendix G, "Fracture Toughness Criteria for Protection Against Failure," to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). For RPV beltline materials having USE below 50 ft-lb, Appendix K, "Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels," to Section XI of the ASME Code provides an equivalent margins analysis (EMA) methodology to demonstrate that adequate fracture toughness is maintained. Appendix H to 10 CFR Part 50 establishes methodologies for determining the increase in transition temperature and the decrease in USE resulting from neutron radiation. GL 92-01, Rev. 1 requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Rev. 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their requirements related to facility RPV material surveillance programs. RG 1.99, Rev. 2, contains RPV integrity evaluations. SRP Section 5.3.2 provides an acceptable method for determining the P-T limit curves for ferritic materials in the beltline of the RPV based on the ASME Code,Section XI, Appendix G methodology. The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2010 Edition of the ASME Code. In addition to use of reference stress intensity factor K1c as fracture toughness and use of a circumferential flaw for circumferential welds, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20% of the preservice hydrostatic test pressure. RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the General Design Criteria (GDC) contained in Appendix A to 10 CFR Part 50. In consideration of the guidance set forth in RG 1.190, GDC 14, "Reactor Coolant Pressure Boundary," GDC 30, "Quality of Reactor Coolant Pressure Boundary," and GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," are applicable. GDC 14 requires the design, fabrication, erection, and testing of the RCPB so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30 requires, among other things that components comprising the RCPB be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31 pertains to the design of the RCPB, stating:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a non brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws. In 10 CFR 50.36, "Technical Specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: ( 1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation is applicable since the amendment revises the ONS TSs for certain P-T limits in TS LCO 3.4.3.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation Enclosure 2 of the February 22,2013, submittal is AREVA Topical Report (TR) ANP-3127, Rev. 1 (the ANP report), "Oconee Nuclear Station Units 1, 2, & 3 Pressure-Temperature Limits at 54 EFPY," January 2013 (ADAMS Accession No. ML 13058A060). The ANP report provides the technical basis for the proposed P-T limits for ONS, Units 1, 2, and 3, based on the TR BAW-1 0046A, Rev. 2 methodology, supplemented by the alternative initial RT NoT of BAW-2308, Revs. 1-A and 2-A. Using this alternative approach is not consistent with 10 CFR Part 50, Appendix G and 10 CFR 50.61, which require a method based on Charpy V-notch (Cv) and drop weight data. Duke Energy hence submitted a request for exemption from the 10 CFR Part 50, Appendix G and 10 CFR 50.61 requirements dated August 3, 2011 (ADAMS Accession No. ML 11223A01 0). This exemption request was approved by the NRC staff on April 26, 2012 (ADAMS Accession No. ML 120580196). The ANP report also contains input parameters such as the initial RT NoT, chemical composition, and adjusted reference temperature (ART) values used in the P-T limits calculations. These material parameters are documented in Table 3-1, Table 3-2, and Table 3-3 in the ANP report, respectively, for ONS, Units 1, 2, and 3. Later, in a supplemental letter dated October 25, 2013, the licensee revised the limiting material information in Table 3-3 of the ANP report, using more appropriate material data for ONS, Unit 3. Detailed information regarding the generation of the P-T limits for each ONS unit in the TR indicated that the proposed P-T limits for each ONS unit consist of the allowable pressures for the controlling RPV beltline, inlet and outlet nozzles, and closure head, with their highest ART values reproduced in Table 1 below. a e e 1g es a ues T bl 1 Th H' h tART V I or e 1m11ng a ena s. F th L' l M t . I Unit Limiting Material ART (°F) 1/4T 3/4T 1 Lower Nozzle Belt Forging 111.9 83.5 1 Intermediate Shell Plate N/A 132.9 1 Upper Shell Longitudinal Weld 171 N/A 1 Intermediate Shell to Upper Shell Circumferential Weld 164.2 132.1 2 Lower Nozzle Belt Forging (Location 3} 161.8 135.7 2 Lower Nozzle Belt Forging (Location 4) 102.4 79.4 2 Upper Shell to Lower Shell Circumferential Weld 193.1 132.5 3 Lower Nozzle Belt Forging (Location 3) 190.8* 160.0* 3 Lower Nozzle Belt Forging {Location 4} 1 06.3* 88.8* 3 Upper Shell to Lower Shell Circumferential Weld 195.6* 162.1 *

• Reflects the rev1sed matenal 1nformat1on 1n the October 25, 2013, supplement. For the thermal analyses, the ANP report adopted a one-dimensional axisymmetric thermal model and used a finite difference technique to develop the temperature distribution. Resulting thermal stresses were then obtained through numerical integration based on the temperature distribution, and the thermal stress intensity factors (K1t) at 1/4T (cooldown) and 3/4T (heatup) locations were developed using the formulas in ASME Code,Section XI, Appendix G. Separately, the membrane stress intensity factor due to unit pressure (Kim) was obtained using the formulas in ASME Code,Section XI, Appendix G for beltline materials. In the final step, the ANP report utilized the applied K1t values and the K1c values at the crack tip to calculate the allowable pressure stress intensity factor (Kip) at the tip of the postulated flaw at the 1/4T and 3/4T locations. Pressure was then obtained by comparing the calculated K1p value to the K1m value. This process was repeated for the closure head and inlet and outlet nozzles, with the K1m formula for nozzles from Welding Research Council (WRC) Bulletin 175, "PVRC [Pressure Vessel Research Committee of the WRC] Recommendations on Toughness Requirements for Ferritic Materials." Based on these different sets of P-T limits, the bounding P-T limits were determined. The resulting P-T limits were further modified to consider minimum boltup temperature and the closure flange limits. 3.2 NRC Staff's Evaluation Neutron Fluence In its February 22, 2013, letter, the licensee stated that" ... projected fluence values at 54 EFPY are based on Topical Report BAW-2241 P-A, Revision 2, which adheres to the guidance contained in NRC Regulatory Guide 1.190, 'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. "' BAW-2241 NP-A, the publicly available version of the methodology, contains a description of the core neutron source and transport calculations used to determine fluence for, among others, Babcock and Wilcox (B&W)-designed nuclear steam supply systems (ADAMS Accession No. ML07331 0660). Revision 2 of this approved TR includes several NRC staff safety evaluations (SEs), all documenting that the approach described in BAW-2241 NP-A is acceptable. Since the licensee's fluence methodology is NRC-approved and applicable to B&W RPVs, the NRC staff determined that the fluence calculations are acceptable for use in determining the ONS, Units, 1, 2, and 3, P-T limits. P-T Limits To verify the Duke Energy calculated ARTs for the limiting materials shown in Table 1 above, the NRC staff performed ART calculations using the materials information in the NRC's Reactor Vessel Integrity Database (RVID) for ONS, Units 1, 2, and 3. For additional assurance, the NRC staff also examined the RPV materials information in the license renewal application (LRA) for the ONS units dated June 1998. The NRC staff found that, except for the initial RT NOT values for Linde 80 welds and some minor discrepancies, the materials information for the RPV materials in the submittal is consistent with the RVID and the 1998 LRA. The neutron fluence values for RPV materials are also close to the values obtained by direct extrapolation of the neutron fluence values in the LRA from 48 EFPY to 54 EFPY. Section 3 of the ANP report presented the RT NOT values of the RPV materials. For the initial RT NoT values for Linde 80 welds, the NRC staff verified that they were determined in accordance with the SEs on BAW-2308, Revisions 1-A and 2-A as permitted by the exemption granted on April 26, 2012. Regarding the discrepancies between the NRC staff's and the licensee's ART values, the NRC staff requested in a request for additional information (RAI) letter dated June 21, 2013 (ADAMS Accession No. ML 13165A147), RAI-2, that the licensee provide the wall thickness for the various RPV materials because the neutron fluence values at 1/4T and 3/4T of the RPV in Table 3-1 of the ANP report appeared to be not based on the same RPV thickness for the same unit. The licensee's response dated September 10, 2013, stated that neutron fluence was calculated using the ONS RPV beltline region wall thickness of 8.44 inches plus cladding thickness. The lower nozzle belt forgings vary in thickness from 8.44 inches to 12 inches. With this information, the NRC staff was able to confirm the licensee's calculated neutron fluence values presented in Table 3-1 of the ANP report. Hence, RAI-2 was resolved. RAI-3 requested justification for the minor revision of the copper value for the lower nozzle belt to the upper shell circumferential weld for ONS, Unit 2, and for the same type of weld for ONS, Unit 3, in Table 3-2 of the ANP report. The licensee's response stated that these updated copper values reflected the additional Linde 80 weld data identified in the 1997 NRC inspection. Hence, RAI-3 was resolved. Regarding the surveillance data, the NRC staff's review of the RPV materials information for ONS, Units 1, 2, and 3, indicated that none of the RPV materials with surveillance data are limiting. Therefore, they have no effect on P-T limits for the three units. Further, the NRC staff's ART calculations, considering the licensee's responses discussed above, confirmed the licensee's identification and ART calculation of the limiting RPV materials for the ONS, Units 1, 2, and 3, under the heatup and cooldown transients. Section 4 of the ANP report presented design basis information for the P-T limits. Section 4.2 of the ANP report states that, "A X. tNs (tNs -the thickness at the nozzle belt) deep corner flaw is postulated on the inside surface of the reactor vessel inlet and outlet nozzles and core flood nozzle corner." This went beyond the TR BAW-10046, Rev. 2, methodology of performing analysis on the RPV outlet nozzle only, and RAI-4 requested for a confirmation of this statement. The licensee's response clarified that actual calculations had been performed on the RPV inlet, outlet, and core flood nozzles to determine the most limiting P-T limits for nozzles in the LAR. Hence, RAI-4 was resolved. Appendix G to 10 CFR Part 50 contains additional requirements for the minimum metal temperature of the closure head flange and vessel flange regions. For RPV closure head limits, Section 4.4 of the ANP report states, "The Pressure-Temperature limits derived for the reactor vessel head-to-flange conservatively bounds the minimum required temperature requirements as given in Table 1 of the Appendix G to 10 CFR Part 50." RAI-5 requested support for this statement. The licensee provided the pressure corrected closure head P-T limits for all three units in its response, which clearly demonstrated that when pressure is above 20% of the pre-service hydrostatic test pressure, the P-T limits derived for the RPV 130 oF), bound the minimum required temperature requirements as given in Table 1 of 10 CFR Part 50, Appendix G (120 oF plus RT Nor of the closure head flange material which is 0 oF for all three ONS units). Hence, RAI-5 was resolved, and the NRC staff verified that the proposed P-T limits as illustrated in Figures 7-1 to 7-9 of the ANP report, have satisfied the additional requirements in 10 CFR Part 50, Appendix G, regarding the minimum metal temperature of the closure head flange and vessel flange regions. For heatup and cooldown transients, Section 4.6 of the ANP report indicated that both ramped and stepped transient definitions are modeled for normal operation heatup and cooldown. RAI-6 requested the licensee to confirm and provide discussion whether the thermal stresses are based on only the stepped transient or the worst cases of the stepped and the ramped transients. The licensee's response clarified that allowable pressures for the stepped transients, the ramped transients, and the steady state responses were compared and limiting values were selected to develop the P-T limits. Hence RAI-6 was resolved. Section 6 of the ANP report discussed pressure corrections using the delta-pressure (b.P) listed in Table 6-1. RAI-7 requested that the licensee clarify whether this b.P was subtracted directly from the calculated pressure values based on the ASME Code,Section XI, Appendix G methodology. RAI-7 also requested that the licensee clarify how the calculated 1/4T metal temperature was adjusted to become the "indicated [reactor coolant system (RCS)] inlet temperature." The licensee's response clarified the b.P subtraction and clarified that the thermal model took care of the temperature difference between the 1 /4T metal temperature and the indicated RCS inlet temperature internally in the computer program. Hence, RAI-7 was clarified. Section 7 of the ANP report presented a summary of results for the ONS, Units 1, 2, and 3, P-T limits at 54 EFPY. Figures 7-1 and 7-2 of this section illustrated the P-T limits for heatup and cooldown, along with pressure-temperature pairs of typical points along the P-T limit curves. RAI-8 requested that the licensee use these pressure-temperature pairs as examples, and provide the corresponding K11s to assist the NRC staff in verifying the proposed P-T limits. The licensee provided the requested K11s to assist the NRC staff's independent calculations discussed below. Hence, RAI-8 was resolved. As mentioned earlier, TR BAW-1 0046A, Rev. 2, was used to generate the P-T limits for each ONS unit. The technical basis for P-T limits is summarized in Section 5.0 of the ANP report, and the results for the ONS, Units 1, 2, and 3, P-T limits at 54 EFPY are summarized in Section 7.0. The NRC staff performed independent P-T limit calculations to verify the P-T limits presented in Figures 7-1 to 7-9 for the three ONS units, considering all relevant information in the ANP report and the September 10, 2013, RAI responses. The NRC staff's calculations considered the pressure correction presented in Section 6.0 of the ANP report and the revised information for ONS, Unit 3, in the October 25, 2013, supplement. The NRC staff's verification indicated that for both heatups and cooldowns and for both high and low pressure, the discrepancies between the NRC staff's and the licensee's calculated temperatures were less than five percent. Considering that the NRC staff used hand calculations in some portions of the verification, the NRC staff concludes that these ANP report determinations are acceptable.

Non-Beltline Materials Not Covered in TR BAW-10046A A generic RAI (RAI-1 ), applicable to ONS, Units 1, 2, and 3, as well as other applicants regarding their LARs related to P-T limits, was issued to Duke Energy by RAIIetter dated June 21, 2013 and is discussed below. The NRC staff expressed two concerns in the generic RAI: (1) RPV nozzles, penetrations, and other discontinuities may exhibit significantly higher stresses than those for the RPV beltline region, potentially resulting in more restrictive P-T limits, even if the RT NDT for these components is lower than that of RPV beltline materials, and (2) ferritic RCPB components that are not part of the RPV may have initial RT NDT values, which may define a more restrictive lowest operating temperature in the P-T limits than those for the RPV beltline materials. Regarding the first concern, the licensee's response indicated that components located above the lower nozzle belt forging to upper shell course circumferential weld are projected to receive neutron fluence less than 5.2x1 016 n/cm2 (E > 1 MeV). Therefore, the P-T curves based on these components followed the methodology of TR BAW-10046A, Rev. 2. The NRC staff finds this response acceptable because, first, the projected 54 EFPY neutron fluence values for the nonbeltline components (e.g., inlet and outlet nozzle forgings) will be well below the threshold for embrittlement assessment. Second, even if the b.RT NDT values were considered, judging from the distance between the heatup and cooldown P-T limits for nozzles and the corresponding composite bounding 32 EFPY P-T limits in TR BAW-10046A, Rev. 2, it is unlikely that the small b.RT NDT for the nozzles would have any impact on the composite bounding P-T limits for EFPYs beyond 32. For components in the lower portion of the RPV, the licensee's response indicated that the projected 54 EFPY neutron fluence values at the lower shell forging to Dutchman forging welds exceed the threshold for embrittlement assessment. However, the ART calculation, using RG 1.99, Rev. 2, showed that the P-T limits for the lower portion of the RPV remained bounded by the closure head flange region of the RPV. The NRC staff finds this response acceptable because the licensee demonstrated that the calculated ART value for the lower portion of the RPV components is not high enough to impact the proposed P-T limits. Regarding the second concern related to ferritic RCPB components that are not part of the RPV, the licensee's proprietary information which was submitted as part of the October 25, 2013, supplemental response, demonstrated that the lowest service temperature (LST) requirements for RCPB (i.e., ASME Code, Section Ill, NB-2332) presented no conflict with the proposed P-T limits based on ASME Code,Section XI, Appendix G. This demonstration is made possible with the following revision of the TR BAW-10046A, Rev. 2, methodology: (1) the conservatism in the LST for the control rod drive mechanism motor tube in BAW-1 0046A, Rev. 2, is removed, and (2) an alternative approach similar to those in ASME Code,Section XI, Appendix G, was performed for the hot leg piping to support the LST requirements of NB-2332. Both revisions are acceptable because additional conservatism is not required by ASME Code,Section XI, Appendix G, and the rigorous alternative approach demonstrated that the proposed P-T limits are more limiting than the hot leg piping P-T limits. Hence, except for the associated USE evaluation for the non-beltline materials, RAI-1 was resolved.

The requirements in 10 CFR Part 50, Appendix G, are applicable for both RPV USE values and P-T limits. When an applicant submits an LAR with RPV neutron fluence values higher than those in the licensing basis (docketed neutron fluence values), it needs to address both the USE and P-T limit evaluations. The NRC staff reviewed the USE values at 48 EFPY for RPV plates and the information related to forging of the ONS units in the Duke Energy 1998 LRA and determined that after considering the neutron fluence increase of approximately 12.5% for the current application, the 54 EFPY USE values for these materials are still above 50 ft-lb, meeting the 10 CFR Part 50, Appendix G requirements. For the ONS RPV welds (all are Linde 80 welds), the NRC staff revisited BAW-2275-A, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY," which was submitted to support the LRAs for B&W plants and performed an independent evaluation as part of its review for this LAR. The BAW-2275-A report used an ASME Code,Section XI, Appendix K-based EMA to demonstrate that the flaw extension and the flaw stability criteria of the elastic plastic fracture mechanics analysis were satisfied and, therefore, adequate fracture toughness was maintained for Linde 80 welds having 48 EFPY USE values below 50 ft-lb. The flaw extension criterion requires that the J-integral resistance (J-R) curve at a flaw extension of 0.1 inch (material resistance) exceed applied J (the driving force) (i.e., J0.1/J > 1) for the subject material. The flaw stability criterion requires that the slope of the applied J is less than the slope of the J-R curve at the intersection of these two curves. The NRC staff's independent evaluation of BAW-2275-A focused on whether this generic EMA for 48 EFPY bounds the limiting ONS Linde 80 weld at 54 EFPY. The NRC staff found that (1) the EMA results for Level A and Level B loadings bound the case for Level C and Level D loadings as indicated by comparing results in Table 5-3 (crack extension) and Figure 5-1 (flaw stability) for Level A and Level B loadings to Figure 6-6 (crack extension and flaw stability) for Level C and Level D loadings, and (2) selection of the bounding Linde 80 weld for the stability analysis was based on Table 5-3 results. Based on these, the NRC staff concluded that the Table 5-3 results alone were sufficient for the independent evaluation. Table 5-3 revealed that J0 1/J is 1.32 for the limiting ONS Linde 80 weld and 1.09 for the bounding Linde 80 weld (both are axial welds). The NRC staff examined Figure 3-1 of BAW-2275-A and noticed that the J0.1 decrease for the bounding Linde 80 weld due to increased neutron fluence is very mild beyond 48 EFPY. Therefore, the margin difference of 23% (i.e., 1.32-1.09) is sufficient to keep the bounding Linde 80 weld for 48 EFPY remain bounding, after considering the loss of J0 1 due to 12.5% increase of neutron fluence for the ONS RPVs. Thus, the NRC staff concludes that the BAW-2275-A EMA conclusions for 48 EFPY still apply to the Linde 80 welds of the ONS units for 54 EFPY. For non-beltline materials having neutron fluence exceeding 1017 n/cm2 (E > 1 MeV), the licensee, in its supplemental response (e-mail) dated December 16, 2013 (ADAMS Accession No. ML 13350A098), reported for each ONS unit: (1) the test-based initial USE value and the 54 EFPY USE value for Dutchman forging, and (2) the generic initial USE value and the 54 EFPY USE value for the lower shell to Dutchman circumferential weld (Linde 80 weld). The NRC staff concludes that item 1 is acceptable because the initial USE values are based on test data; and the calculated 54 EFPY USE values are above 50 ft-lb, satisfying the 10 CFR Part 50, Appendix G requirement. The NRC staff concludes that item 2, however, is not acceptable because the NRC staff has not approved a generic initial USE value for Linde 80 welds (see SE related to the LRA for Davis-Besse Nuclear Power Station, ADAMS Accession No. ML 13248A267). Again, using an EMA is the right approach. Since the information needed for this EMA is implicitly in Table 5-3 of BAW-2275-A, the NRC staff performed an independent evaluation instead of issuing another follow-up RAI. Due to the low neutron fluence and the circumferential orientation of the lower shell to Dutchman circumferential weld, the NRC staff estimated that the associated J0 1/J value for this weld would be at least 3.0, judging from the Jo.1/J values for other ONS Linde 80 circumferential welds. Therefore, the NRC staff determined that the BAW-2275-A EMA conclusions for 48 EFPY apply to the non-beltline Linde 80 welds of the ONS units for 54 EFPY. In summary, the NRC staff concludes that all beltline and non-beltline RPV materials of the ONS units meet the 10 CFR Part 50, Appendix G, USE requirements. Based on the above evaluation, the NRC staff determined that the licensee's proposed P-T limits are in accordance with TR BAW-10046A, Rev. 2, with the revision as stated above, and satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50, as modified by the exemption granted on April 26, 2012. Hence, the NRC staff concludes that the licensee's proposed P-T limits, valid for 54 EFPYs, are acceptable for operation of the ONS, Units 1, 2, and 3, RPVs. 3.3 Conclusion Based on the NRC staff's review of the information provided in the licensee's February 22, September 10, October 25, November 29, and December 16, 2013, submittals, the NRC staff concludes that the proposed ONS RPV P-T limits, valid for 54 EFPYs, meet the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50, as modified by the associated exemption granted on April 26, 2012. The NRC staff's conclusion is based on several independent NRC staff evaluations and its verification that the proposed P-T limits were developed appropriately using the TR BAW-10046A, Rev. 2, methodology, with the revision as stated in Section 3.2 of this SE. Further, the NRC staff determined that the ONS RPV USEs also meet the requirements of Appendix G to 10 CFR Part 50. Therefore, the NRC staff concludes that the licensee's proposed TS revisions to reflect the use of these P-T limits are appropriate.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on April 16, 2013 (78 FR 22568). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: Date: February 27, 2014 S. Sheng B. Parks S. Batson If you have any questions, please call me at 301-415-1030. Sincerely, Ira/ Richard V. Guzman, Senior Project Manager Plant Licensing Branch 11-1 Docket Nos. 50-269, 50-270, and 50-287

Enclosures:

1. Amendment No. 384 to DPR-38 2. Amendment No. 386 to DPR-47 3. Amendment No. 385 to DPR-55 4. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPLII-1 R/F RidsAcrsAcnw_MaiiCtr Resource RidsNrrDoriDpr Resource RidsNrrDoriLp2-1 Resource RidsNrrDeEvib Resource RidsNrrDssSrxb Resource RidsNrrDssStsb Resource RidsNrrLASFigueroa Resource RidsNrrPMOconee Resource RidsRgn2MaiiCenter Resource SSheng SParks RGrover Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ADAMS Accession No.: ML 14041A093 *See memo dated 1/23/14 OFFICE NRRILPL2-1/PM NRRILPL2-1/LA NRRIEVIB/BC* NRRISTSB/BC OGC NRRILPL2-1/BC NRRILPL2-1/PM NAME RGuzman SFigueroa SRosenberg REIIiott JWachutka RPascarelli RGuzman DATE 2/12/14 2/11/14 1/23/14 2/26/14 2/18/14 2/27/14 2/27/14 OFFICIAL RECORD COPY